ML20321A106
| ML20321A106 | |
| Person / Time | |
|---|---|
| Issue date: | 04/30/2020 |
| From: | Office of Nuclear Material Safety and Safeguards |
| To: | |
| AWillis - 301.415.0479 ; SKumar | |
| Shared Package | |
| ML20321A086 | List:
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| References | |
| NUREG-2215 | |
| Download: ML20321A106 (26) | |
Text
11-1 OPERATION PROCEDURES AND SYSTEMS EVALUATION Review Objective The objective of the U.S. Nuclear Regulatory Commissions (NRCs) review of operating procedures and operations systems for a dry storage system (DSS) or dry storage facility (DSF) is to evaluate associated applications for clarity and completeness to verify the following:
The description of the applications provides sufficient detail to ensure that reviewers can understand the operations and their effects on the design evaluations.
The DSS or DSF operations are consistent with the design bases for which the DSS or DSF was designed and analyzed in the other chapters of the safety analysis report (SAR) and this standard review plan (SRP).
The DSS or DSF operations incorporate and are consistent with the conditions of the certificate of compliance (CoC) or a specific license, including the proposed technical specifications.
Applicability This chapter applies to the review of applications for specific licenses for an independent spent fuel storage installation (ISFSI) or a monitored retrievable storage installation (MRS), categorized as a DSF. This chapter also applies to the review of applications for a CoC of a DSS for use at a general license facility. Sections or tables of this chapter that apply only to a DSF specific license application (for an ISFSI and MRS) are identified with (SL) in the heading. Sections or tables that apply only to DSS CoC applications have (CoC) in the heading. A subsection without an identifier applies to both types of application.
Areas of Review This chapter addresses the following areas of review:
operation description storage container loading storage container handling and storage operations storage container unloading repair and maintenance (SL) other operating systems (SL) operation support systems (SL) control room and control area (SL) analytical sampling (SL) fire and explosion protection (SL)
Regulatory Requirements and Acceptance Criteria This section summarizes those parts of Title 10 of the Code of Federal Regulations (10 CFR) Part 72, Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater than Class C Waste, that are relevant to the review areas this chapter addresses. The NRC staff reviewer should refer to the
11-2 exact language in the regulations. Tables 11-1a and 11-1b match the relevant regulatory requirements to the areas of review covered in this chapter for specific license and CoC reviews, respectively. In addition, requirements in 10 CFR Part 20, Standards for Protection Against Radiation, also apply to reviews for specific license applications. The reviewer should coordinate with the radiation protection reviewer (Chapter 10A of this SRP) to determine the applicable 10 CFR Part 20 requirements.
11-3 Table 11-1a Relationship of Regulations and Areas of Review for a DSF (SL)
Area of Review 10 CFR Part 72 Regulations 72.24 (b)(e)(f)(l) 72.40 (a)(5)(13) 72.44 (c)(1)(2)(3)(5),
(d)(1)(2) 72.104 72.106(b) 72.122 (f)(h)(i)(j)(k)(l) 72.124 72.126 (a)(2)(3)(4),
(b)(c)(d) 72.128 (a)(1)(2) 72.150 Operation Description Storage Container Loading Storage Container Handling and Storage Operations Storage Container Unloading Repair and Maintenance Other Operating Systems Operation Support Systems Control Room and Control Area Analytical Sampling Fire and Explosion Protection Table 11-1b Relationship of Regulations and Areas of Review for a DSS (CoC)
Area of Review 10 CFR Part 72 Regulations 72.104(b) 72.106(b) 72.124 72.234(f) 72.236(c)(d)(f)(g)(h)(i)(l)(m)
Operation Description Storage Container Loading Storage Container Handling and Storage Operations Storage Container Unloading
11-4 The following sections describe acceptance criteria, which are designed to ensure that the applicant fully describes the information on systems and significant operating sequences and actions in the SAR chapters. A sufficient level of detail is needed for the reviewer to conclude that the DSS or DSF operations are consistent with the design bases, will adequately protect health and minimize danger to life or property, protect the fuel from significant damage or degradation, and provide for the safe performance of tasks and operations. The applicant should provide an adequate description of the functional systems operations and identify the proper functioning of each system in a manner that adequately supports the purposes described above for the operations procedures descriptions and the evaluations in the other chapters of the SAR.
11.4.1 Operation Description Operation description relates to the overall storage functions and operation of the DSS or DSF.
The description should identify and describe the sequences of operations, actions, and controls that are important to safety for spent nuclear fuel (SNF), high-level radioactive waste (HLW), and reactor-related greater-than-Class-C (GTCC) waste handling and storage, including loading and unloading operations, as applicable. Sufficient detail should be included to enable the reviewer to evaluate engineering and operational controls. The operation description also should include the principal design features, procedures, and special techniques associated with criticality prevention, chemical safety, operation shutdown modes, instrumentation, radiation protection, protection of radioactive contents from significant damage or degradation, and maintenance techniques. The description should be sufficiently detailed to provide for the safe performance of tasks and operations.
Major operating procedures should exist for the principal activities expected to occur during loading, storage preparation, dry storage, and unloading. Section 11.3 above describes the areas of review for the SAR operating procedure descriptions, as does Chapter 8 of Regulatory Guide (RG) 3.61, Standard Format and Content for a Topical Safety Analysis Report for a Spent Fuel Dry Storage Cask, relevant sections of RG 3.48, "Standard Format and Content for the Safety Analysis Report for an Independent Spent Fuel Storage Installation or Monitored Retrievable Storage Installation (Dry Storage), " and RG 3.62, Standard Format and Content for the Safety Analysis Report for Onsite Storage of Spent Fuel Storage Casks. The applicant should submit operating procedure descriptions as part of the application to address the DSS or DSF design features and operations.
Operating procedure descriptions should identify measures to control processes and mitigate potential hazards that may be present during planned, normal operations.
Section 11.5 of this chapter discusses the review of previously identified processes and potential hazards.
Operating procedure descriptions should ensure conformance with the applicable operating controls and limits described in the DSS CoC or DSF license conditions and technical specifications provided in the SAR chapter on technical specifications and operating controls and limits.
Operating procedure descriptions should reflect planning to ensure that operations will fulfill the following acceptance criteria:
Occupational radiation exposures will be maintained as low as is reasonably achievable (ALARA) and within the limits of 10 CFR Part 20.
11-5 Effective measures will be taken to preclude potential unplanned and uncontrolled releases of radioactive materials and otherwise minimize potential releases under normal operations conditions.
Doses for members of the public will be maintained within the limits of 10 CFR Part 20 and 10 CFR 72.104, Criteria for radioactive materials in effluents and direct radiation from an ISFSI or MRS, for normal operations, and 10 CFR 72.106, Controlled area of an ISFSI or MRS, for accident conditions.
In addition, the operating procedure descriptions should support and be consistent with the bases used to estimate radiation exposures and total doses as defined in the radiation protection review guidance in this SRP that applies to the particular application (Chapter 10A for specific license applications and Chapter 10B for CoC applications).
Operating procedure descriptions should include provisions for the following activities:
testing, surveillance, and monitoring of the stored material and storage containers during storage and loading and unloading operations contingency actions triggered by inspections, checks, observations, instrument readings, and so forth; some of these may involve off-normal and accident conditions addressed in the chapter of the SAR on accident analyses RG 3.61, RG 3.62, and RG 3.48 provide further detail on operating procedure descriptions.
11.4.2 Storage Container Loading In addition to the acceptance criteria specified above for the operation description, there are additional acceptance criteria for storage container loading, as follows:
The operating procedures descriptions should include provisions for loading of SNF, reactor-related GTCC waste, and HLW storage containers, as applicable.
The operating procedure descriptions should facilitate reducing the amount of water vapor and oxidizing material within the storage container to an acceptable level in order to protect the SNF cladding against degradation that might otherwise lead to gross ruptures.
Operating procedures should specify methods for placing damaged fuel in a damaged-fuel can before loading into a SNF storage container, as applicable.
11.4.3 Storage Container Handling and Storage Operations The regulatory requirements in 10 CFR 72.24, Contents of application: Technical information, (SL), 10 CFR 72.124, Criteria for nuclear criticality safety, 10 CFR 72.128, Criteria for spent fuel, high-level radioactive waste, and other radioactive waste storage and handling, (SL), and 10 CFR 72.236, Specific requirements for spent fuel storage cask approval and fabrication, (CoC) address the information to be included in a SAR for handling storage containers loaded with SNF, HLW, and reactor-related GTCC waste, as applicable to review of CoC and specific license applications. The SAR should include information as described in RG 3.61,RG 3.62, and RG 3.48 on handling systems for SNF and reactor-related GTCC waste (and HLW if for a MRS),
11-6 as applicable. The descriptions of the SNF, HLW, or reactor-related GTCC waste handling systems and operations should be clear. The applicant should address the functions of transfer from transportation vehicles, receipt inspection, and initial decontamination. The applicant should include a statement indicating whether the NRC reviewed these operations or the systems used to perform these operations, as applicable, as part of a licensing action under 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, or 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants. The SAR should include a description of the transfer facility and its use, including its use during the stages of operation of the DSS or DSF.
The descriptions should include consideration of potential off-gassing (including hydrogen generation).
11.4.4 Storage Container Unloading In addition to the acceptance criteria specified above for the operation description, the descriptions should include provisions for unloading SNF, reactor-related GTCC waste, and HLW, as applicable. The operating procedures should facilitate ready retrieval of the contents stored in the DSS or DSF storage containers.
11.4.5 Repair and Maintenance (SL)
The SAR should contain a description of the repair and maintenance facilities and describe the operation of these facilities, including provision for contamination control and occupational exposure minimization. Chapter 12, Conduct of Operations Evaluation, of this SRP provides useful guidance for the evaluation of maintenance operations. Note that the maintenance and use of a transportation package for shipping radioactive material is governed only by the requirements in 10 CFR Part 71, Packaging and Transportation of Radioactive Material. Thus, any repair and maintenance operations involving transport packaging must be done in accordance with those requirements (see 10 CFR 71.17, General License: NRC-Approved Package, and 10 CFR Part 71, Subpart H, Quality Assurance), including requirements in the transport CoC for the package.
11.4.6 Other Operating Systems (SL)
The scope of the review of this section includes all operating systems important to safety that are not covered in the preceding sections, except for the acceptance criteria for instrumentation and control (I&C), which are in Section 11.4.7, Operation Support Systems, and acceptance criteria for analytical sampling, which are in Section 11.4.9, Analytical Sampling, of this SRP. The applicant should prepare the SAR sections on auxiliary systems and other operating systems that are important to safety as described in RG 3.62 and RG 3.48 and noted in the narrative descriptions or flowcharts describing the operation of the ISFSI or MRS. The regulations in 10 CFR 72.24 require that the SAR include clear descriptions of the systems and system equipment and controls used to assure safety. These items should be consistent with other parts of the SAR. Examples of other operating systems that may be classified as important to safety include ventilation and off-gas systems, electrical systems, air supply systems, steam supply and distribution systems, water supply systems, fire protection systems, air sampling systems, decontamination systems, and systems related to chemical hazards. This information should include an analysis or other acceptable basis for determining that operation support systems important to safety remain operational under off-normal and accident conditions.
11-7 11.4.7 Operation Support Systems (SL)
The SAR should include information on operation support systems, primarily I&C systems and component spares or alternative equipment, as required by 10 CFR 72.122, Overall requirements. These items should be as described in RG 3.62 and RG 3.48. This information should include an analysis or other acceptable basis for determination that operation support systems important to safety remain operational under normal, off-normal, and accident conditions.
The SAR should include clear descriptions of the operation support systems and descriptions of equipment and controls used to assure safety and that are consistent with other parts of the SAR.
11.4.8 Control Room and Control Area (SL)
The SAR should include a discussion of how a control room and control areas permit the installation to operate safely under normal, off-normal, and accident conditions (10 CFR 72.122(j)
The SAR should include clear descriptions of the control room and control area. In addition, 10 CFR 72.122(j) requires that a control room or control area, if appropriate for the DSF design, must be designed to permit occupancy and actions to be taken to monitor the safety of the DSF under normal conditions and to provide safe control of the DSF under off-normal or accident conditions.
The NRC has accepted omission of a control room for ISFSI or MRS operations that have not involved use of a powered cooling system for material in storage.
11.4.9 Analytical Sampling (SL)
The SAR should include a discussion of the provisions for obtaining samples for analyses necessary to ensure that the ISFSI or MRS is operating within prescribed limits. The SAR should include a description of the facilities and equipment available to perform the required tests.
11.4.10 Fire and Explosion Protection (SL)
The regulations in 10 CFR 72.122(c) require the DSS or DSF structures, systems, and components (SSCs) important to safety and their contents to have adequate protection against fires and explosions to ensure the SSCs continue to effectively perform their safety functions under credible or design-basis fire and explosion conditions.
The regulations in 10 CFR 72.122(c) require the applicant to take measures for fire prevention, fire detection, fire suppression, and fire containment for the protection of the DSS or DSF SSCs important to safety and their contents. The SAR should include a discussion of these capabilities.
Review Procedures The focus of this review is twofold: (1) the operations descriptions of the DSS or DSF and (2) the functions needed for operability and the compatibility of proposed systems with performance of those functions. The NRC does not review and approve detailed procedures (e.g., standard operating procedures). However, the NRC does review, and the SAR should include, operations procedure descriptions that are sufficient in detail to illustrate the important actions and processes to be done and demonstrate that the operations will be conducted in a manner that (1) is consistent with the CoC or license conditions, as appropriate, and technical specifications, (2) ensures that the DSS or DSF operations will be consistent with the design bases and fulfill safety functions, and (3) includes adequate consideration of radiation protection and ALARA for the public and personnel. Also, the review of the descriptions of functions of the proposed
11-8 systems constitutes another principal basis for assessing that the DSS or DSF will be operated in the manner described above. Reviews in other SRP sections determine quantitative functional performance for functional and structural performances.
Figure 11-1 shows the interrelationship between the operating procedures evaluation and the other areas of review described in this SRP.
An applicants operating procedures are, in a significant way, how the applicants conduct of operations is implemented. Therefore, the review should be coordinated with the conduct of operations review (SRP Chapter 12) to ensure that there are no inconsistencies.
11-9 Figure 11-1 Overview of Operation Procedures and System Evaluation Chapter 11 - Operation Procedures and Systems Evaluation Operation Description Operation Support Systems (SL)
Storage Container Loading Control Room and Control Area (SL)
Storage Container Unloading Analytical Sampling (SL)
Storage Container Handling and Storage Operations Repair and Maintenance (SL)
Other Operating Systems (SL)
Fire and Explosion Protection (SL)
Chapter 4 -
Structural Evaluation
- Description of the SSCs
- Design Criteria
- Loads
- Normal and Off-normal Conditions
- Accident Conditions Chapter 5 -
Thermal Evaluation
- Decay Heat Removal System
- Material and Design Limits
- Thermal Loads and Environmental Conditions Chapter 7 - Criticality Evaluation
- Criticality Design Criteria and Features
- Fuel Specification
- Model Specification
- Criticality Analysis
- Burnup Credit
Chapter 8 - Materials Evaluation
- Material Properties
- Environmental Degradation; Chemical and Other Reactions
- Fuel Cladding Integrity and Retrievability
- Code Use and Quality Standards Chapter 9 -
Confinement Evaluation
- Confinement Monitoring Capability Chapter 6 -
Shielding Evaluation
- Shielding Design Description
- Radiation Source Definition
- Shielding Model Specification
- Shielding Analyses
Chapter 12 -
Conduct of Operations Evaluation
- Acceptance Tests
- Preoperational Testing and Startup Operations (SL)
- Normal Operations (SL)
- Maintenance Program Chapter 3 -
Principal Design Criteria
- Design Bases for SSCs Important to Safety
- Design Criteria for Safety Protection Systems
Chapter 10A (SL)/10B (CoC) - Radiation Protection Evaluation
- Radiation Protection Design Features
- Dose Assessment
- Health Physics Program (SL)
Chapter 17 - Technical Specifications Evaluation
- Functional and Operating Limits, Monitoring Instruments, and Limiting Control
- Monitoring Instruments
- Surveillance Requirements
- Administrative Controls Chapter 15 - Quality Assurance Evaluation
- Design and Nonconformance
- Procurement and Test Control Chapter 15 -
Quality Assurance Evaluation
- Design and Nonconformance
- Procurement and Test Control Chapter 16 -
Accident Analysis Evaluation
- Corrective Course of Action Chapter 1 -
General Information
- Site Description (SL)
- DSS or DSF Description and Operational Features
- Engineering Drawings
- Contents
- Quality Assurance Program (SL)
Chapter 17 -
Technical Specifications Evaluation
- Functional and Operating Limits, Monitoring Instruments, and Limiting Control
- Surveillance Requirements
- Administrative Controls Chapter 14 -
Decommissioning Evaluation (SL)
- Proposed Decommissioning Plan
- Operational Features Chapter 13 - Waste Management Evaluation (SL)
- Off-Gas Treatment and Ventilation
- Liquid Waste Treatment and Retention
- Solid Wastes Chapter 13 -
Waste Management Evaluation (SL)
- Waste Sources and Waste Management Facilities
- Off-Gas Treatment and Ventilation
- Liquid Waste Treatment and Retention
- Solid Wastes
11-10 11.5.1 Operation Description Review the description of operation systems functions for completeness. Compare the functions with descriptions included in other licensing documentation to confirm acceptability. For a specific license application, if a previously certified DSS design is used, check the functions described in the DSF SAR under review for compatibility with those functions that were included in the SAR for the certified DSS.
Review flowcharts and narrative descriptions of steps as provided on general operating functions.
Ensure that the applicant has adequately described the appropriate operations, equipment involved, and personnel requirements.
Review the operating procedure sequences described in the SAR. Use the direct dose rate information in the chapter of the SAR on shielding to assess compliance with radiation protection requirements. Coordinate the evaluation of the operating procedure sequences with the shielding and radiation protection evaluations covered in Chapter 6, Shielding Evaluation, and Chapter 10A (for DSFs) or 10B (for DSSs) of this SRP.
American National Standards Institute (ANSI)/American Nuclear Society (ANS) 57.9, Design Criteria for an Independent Spent Fuel Storage Installation (Dry Storage Type), applies to dry storage operating procedures. NUREG/CR-4775, Guide for Preparing Operating Procedures for Shipping Packages, issued December 1988, provides guidance on preparing operating procedures for shipping packages. Although NUREG/CR-4775 specifically addresses 10 CFR Part 71, most of the guidance can be adapted for storage casks that are governed by 10 CFR Part 72. Therefore, become familiar with this information before initiating the operating procedures review.
The DSF applicant will, as the DSF licensee, develop detailed written procedures (e.g., standard operating procedures) that should be based on the operations descriptions in the SAR. For DSSs, the licensee that will use the DSS will develop the detailed procedures. These detailed procedures should be based on the operations descriptions in the DSS SAR operations chapter.
The reviewer should ensure that the CoC contains a condition that makes this a requirement. In general, perform the following actions in the process of evaluating all of the operating procedure descriptions and operational sequences provided in the SAR:
Verify that the proposed operating procedure descriptions incorporate and are compatible with the applicable operating limits and controls in the chapter of the SAR on technical specifications and operational controls and limits. Coordinate with the operating controls and limits review, as described in Chapter 17, Technical Specifications Evaluation, of this SRP.
Ensure that the proposed operating procedure descriptions properly consider the prevention of hydrogen gas generation from any cause. Prevention of hydrogen generation or adequate purging of hydrogen is essential during loading and unloading operations that involve seal welding, seal cutting, grinding, or other forms of hot work.
Determine whether the descriptions include appropriate precautions to minimize occupational radiation exposures in accordance with ALARA principles and the limits given in 10 CFR Part 20, as required in 10 CFR 72.24(e), and consistent with the requirements in 10 CFR 72.126(a). Provisions may include the use of remotely controlled equipment, monitoring, and the use of portable shielding.
11-11 Verify that the operating procedure descriptions include a general listing of the major tools and equipment needed to support loading, preparation for storage, storage, and unloading operations. Confirm that the descriptions address installation, use, and removal of the storage container and its contents, tools, and equipment. In addition, ensure that the descriptions address any specialized tools and equipment, such as lifting yokes, transporter equipment, welding and cutting equipment, and vacuum drying equipment, in sufficient detail to provide a clear understanding of their function(s). The use of any such equipment is subject to approval as part of the application review if that equipment is either classified as being important to safety or, though not important to safety, per the design bases, the equipments failure could negatively impact fulfillment of a function that is important to safety. Ensure that the SAR identifies and describes such equipment in detail, identifies the performance characteristics of the equipment, and contains an evaluation the equipments design.
11.5.2 Storage Container Loading The operating procedure descriptions in the SAR should present the activities sequentially in the anticipated order of performance. Review the generic procedures in the SAR to ensure that they include appropriate key prerequisite, preparation, and receipt inspection activities to be accomplished before storage container loading. Verify that the SAR specifies the tests, inspections, verifications, and cleaning procedures required in preparation for storage container loading. In addition, where applicable, verify that the procedure descriptions include actions needed to ensure that any fluids such as shield water and primary coolants fill their respective cavities according to design specifications. In addition, verify that the procedure descriptions incorporate the applicable operating controls and limits described in the chapter of the SAR on technical specifications and operating controls and limits. These controls and limits include any dose rate and contamination measurements necessary to confirm compliance with the respective limits in the technical specifications.
Specifications for Spent Nuclear Fuel, Reactor-Related Greater-Than-Class-C Waste, and High-Level Radioactive Waste Verify that the loading procedure description appropriately addresses the SNF specifications (e.g., burnup, cooling period, source terms, heat generation, cladding damage, associated nonfuel hardware) in the chapters of the SAR on principal design criteria and technical specifications and operation controls and limits. For storage containers relying on burnup credit, ensure that the loading procedure description includes verification that assemblies selected for loading meet the specifications for assembly operational history and the burnup credit loading curve. In addition, ensure that the loading procedure description includes performance of measurements to confirm assembly burnup values. For general license facilities and for specific license DSFs used to store SNF from a co-located 10 CFR Part 50 or 10 CFR Part 52 reactor facilitys SNF pool, depending on the types and specifications of fuel assemblies stored in the reactor SNF pool, detailed site-specific procedures may be necessary to ensure that all fuel loaded in the storage container meets the fuel specifications for the storage container design. These detailed procedures can be evaluated only on a site-specific basis and will generally be evaluated through inspections rather than during the licensing review. However, check that the SAR indicates that such procedures may be necessary and describes the essential elements of the procedures.
(SL) For specific license DSFs that will also store reactor-related GTCC waste or MRSs that will store HLW, verify that the loading procedure description appropriately addresses the waste specifications and the acceptance criteria for storage at the facility that are described in the SARs
11-12 principle design criteria and technical specification and operation controls and limits chapters. For DSFs that receive SNF, reactor-related GTCC waste, or HLW from other locations (besides the 10 CFR Part 50 and 10 CFR Part 52 waste with which the DSF may be co-located), ensure that the operations descriptions include how the licensee will ensure the items received at the DSF meet the license specifications for storage at the DSF.
Damaged Fuel Verify that the SAR includes appropriate measures for the loading of damaged fuel, if damaged fuel is included in the proposed storage container contents. Chapter 3, Principal Design Criteria Evaluation, and Chapter 8, Materials Evaluation, of this SRP provide criteria for the storage of damaged fuel. Use information in Section 8.5.15.1, Spent Fuel Classification, of this SRP to identify the conditions that determine when SNF is to be classified as damaged fuel. Review Sections 8.5.15.1 and 8.5.15.3 of this SRP to determine the classification, documentation, and special confinement requirements for damaged fuel, and determine whether operating procedures address these requirements.
Subcriticality Features Where applicable, verify that the procedure descriptions include the use of temporary or removable features important to criticality safety that may require installation during loading operations. Such items include fuel spacers and items (e.g., blocks) used to prevent loading of contents in selected SNF basket locations. The procedure descriptions should include installation, or verification of the installation, of these items before loading for storage containers that rely upon these features in the criticality analysis. Additionally, ensure that the procedure descriptions include verification, in accordance with technical specification requirements, of the minimum soluble boron level necessary for SNF loading into storage containers that require soluble boron to ensure subcriticality.
ALARA Verify that the procedure descriptions incorporate ALARA principles and practices. These may include provisions to perform radiological surveys, establish exposure and contamination control measures, and use or install temporary shielding and inclusion of caution statements related to actions that could change radiological conditions.
Offsite Release Where applicable, verify that the SAR describes methods to minimize offsite releases. Examples of these methods include, but are not limited to, decontamination of the storage containers, means for minimizing contamination of DSS or DSF SSCs, controls for processing of liquids and gases removed from the storage container during the draining and drying process, filtered ventilation, and temporary containments (tents). Ensure that the procedure descriptions also provide for minimizing the generation of radioactive waste.
Draining and Drying Evaluate the descriptions related to methods for use in draining and drying the storage container for wet loading operations and, if applicable, HLW and reactor-related GTCC waste containers.
Ensure that the SAR clearly describes the procedures for removing water vapor and oxidizing material to an acceptable level. Assess whether those procedures are appropriate.
11-13 The NRC staff has accepted vacuum drying methods comparable to those recommended in PNL-6365, Evaluation of Cover Gas Impurities and Their Effects on the Dry Storage of LWR Spent Fuel, issued November 1987 (Knoll and Gilbert 1987). This report evaluates the effects of oxidizing impurities on the dry storage of light-water reactor (LWR) fuel and recommends limiting the maximum quantity of oxidizing gasses (such as oxygen, carbon dioxide,1 and carbon monoxide) to a total of 1 gram-mole per cask. This corresponds to a concentration of 0.25 volume percent of the total gases for a 7.0-cubic-meter (about 247-cubic-foot) container gas volume at a pressure of about 0.15 megapascals (MPa) (1.5 atmosphere) at 300 Kelvin (K) (80.3 degrees Fahrenheit (°F)). This 1-gram-mole limit reduces the amount of oxidants below levels where any cladding degradation is expected. Moisture removal is inherent in the vacuum drying process, and levels at or below those evaluated in PNL-6365 (about 0.43 gram-mole water) are expected if adequate vacuum drying is performed.
If alternative methods other than vacuum drying are used (such as forced helium recirculation),
ensure that the applicant provides additional analyses or tests to sufficiently justify that cover gas moisture and impurity levels as specified in the chapter of the SAR on operating procedures are met and will not result in unacceptable cladding degradation.
The following examples illustrate the accepted methods for container draining and drying in accordance with the recommendations of PNL-6365:
The container should be drained of as much water as practicable and evacuated to less than or equal to 4.0x10-4 MPa (4 millibar, 3.0 millimeters of mercury or Torr). After evacuation, adequate moisture removal should be verified by maintaining a constant pressure over a period of about 30 minutes without vacuum pump operation (or the vacuum pump is running but is isolated from the container with its suction vented to atmosphere). The container is then backfilled with an inert gas (e.g., helium) for applicable pressure and leak testing, with care being taken to preserve the purity of the cover gas. After backfilling, cover gas purity should be verified by sampling.
The procedures should reflect the potential for blockage of the evacuation system or masking of defects in the cladding of nonintact rods for SNF storage containers as a result of icing during evacuation. Icing can occur from the cooling effects of water vaporization and system depressurization during evacuation. Icing is more likely to occur in the evacuation system lines than in the container because of decay heat from the fuel. A staged drawdown or other means of preventing ice blockage of the container evacuation path may be used (e.g., measurement of container pressure not involving the line through which the container is evacuated).
The procedures should specify a suitable inert cover gas (such as helium) with a quality specification that ensures a known maximum percentage of impurities to minimize the source of potentially oxidizing impurity gases and vapors and adequately remove contaminants from the container.
The process should provide for repetition of the evacuation and repressurization cycles if the container interior is opened to an oxidizing atmosphere following the evacuation and repressurization cycles (as may occur in conjunction with remedial welding, seal repairs).
1 Can be broken down by radiolysis.
11-14 Ensure that the drying specifications are consistent with the proposed operating controls and limits described in the technical specifications provided in the SAR. In addition, assess the need for any additional technical specifications.
Welding and Sealing Coordinate with the structural and materials reviewers evaluation of welded lids as described in Section 8.5.9, Bolt Applications, of this SRP for applying the proper weld joint, welding procedures, and nondestructive examination methods to ensure that the appropriate operating procedures are in place and acceptable. Verify that the procedures are acceptable for nondestructive examination and welding of the closure welds. Confirm that the SAR also ensures that ALARA principles are followed and includes acceptable provisions for correcting weld defects and any additional drying and purging that may be necessary.
Verify that provisions for placing and tightening any closure bolts, such as those associated with concrete overpacks, are consistent with information presented in SAR chapters that address applicable design criteria, structural evaluation, and the acceptance tests and maintenance program. The materials discipline should ensure that the closure bolts satisfy the conditions given in Section 8.5.9, Bolt Applications, of this SRP. Ensure that the SAR specifies the torque required to properly seal the closure lid. The inner seal should be tested using a helium leak test with the interior of the cask pressurized as previously described. The outer seal should also be tested using a helium leak test with the between-seal volume pressurized as required by the respective subsection of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section III, Rules for Construction of Nuclear Facility Components.
Filling and Pressurization Verify that the procedure recommendations address steps to fill and pressurize the confinement with inert gas such as helium with a known maximum percentage of impurities. The operating procedures should state that the filling and pressurization (or evacuation and backfill) process be repeated if the cavity is exposed to the atmosphere. Ensure that the procedure recommendations include the requirements in the chapter of the SAR on technical specifications and operation controls and limits.
Ensure that the SAR specifies the leak rate criteria (e.g., total leakage, leakage per closure, sensitivities of tests). Verify that these criteria are consistent with those presented in chapters of the SAR on principal design criteria, operating procedures, and technical specifications and operational controls and limits. In addition, assess the general methods of leak testing (e.g., pressure rise, mass spectrometry) as they apply to the leak rate being tested. Pay particular attention to the possible use of quick-disconnect fittings for draining and filling operations.
Although no credit is taken for these devices as part of the confinement boundary, if leaking, their presence can cause leak paths through adjacent welds and affect the results of the leak test; the SAR should provide guidance regarding their use. In addition, the guidelines presented in the SAR should note that leak testing is in accordance with ANSI N14.5, Radioactive Materials Leakage Tests on Packages for Shipment.
Ensure that the SAR presents applicable pressure testing criteria (e.g., test pressure, hold periods, inspections).
11-15 11.5.3 Storage Container Handling and Storage Operations For sites that propose to have SNF, HLW, or reactor-related GTCC waste handling facilities, the SAR should include operations descriptions for some or all of the following functions:
receipt and inspection of loaded transportation packages SNF, HLW, or reactor-related GTCC waste transfer and examination fuel reconstitution SNF, HLW, or reactor-related GTCC waste container short-term storage storage container decontamination storage container loading and storage preparation storage container transfer to storage SNF, HLW, or reactor-related GTCC waste container removal from the storage pad and transfer into a transport package damaged fuel element containerization Ensure that the applicant has adequately described the appropriate procedures, equipment involved, and personnel requirements. Ensure that the receipt, handling, and transfer descriptions include a functional description of the associated systems and descriptions of all features, systems, or special handling techniques that provide for safe operations under both normal and off-normal conditions. Follow the same procedures to review operations for handling HLW or reactor-related GTCC waste, as applicable, as used for SNF. Because the SNF, HLW, and reactor-related GTCC waste handling systems have many interfaces with other systems of the facility (e.g., SNF pool), verify that the applicant addressed these interfaces and that continuity of operations can occur under all operational conditions.
Pay particular attention to ensure that all accident events applicable to transfers are bounded by the design events analyzed in the chapters of the SAR on principal design criteria, structural evaluation, and accident analyses. This includes procedures to be specified in the SAR for use after a design-basis accident for testing the effectiveness of the shielding. The structural and thermal disciplines should coordinate their review to ensure that all conditions for lifting and handling methods are bounded by the evaluations in their respective chapters of the SAR.
Coordinate as needed with the review of Chapter 17 of this SRP regarding technical specifications associated with cask transfer operations, such as restricting lift heights and environmental conditions (e.g., high and low temperatures).
Verify that the necessary operations descriptions include inspection, surveillance, and maintenance activities that are applicable during storage. Ensure that the appropriate surveillance and monitoring requirements are also include in the chapter of the SAR on technical specifications and operational controls and limits, and that maintenance is included in the chapter of the SAR on conduct of operations. Note that if the confinement vessel closure is bolted, the
11-16 NRC generally requires that the successful operation of the seals be demonstrated with an initial leak test and a monitoring system or a surveillance program, or both, as discussed in Chapter 12 of this SRP.
The shielding and radiation protection reviewers should verify that proposed operations descriptions give due consideration to maintaining ALARA with respect to doses during storage container handling and storage operations.
11.5.4 Storage Container Unloading Verify that the SAR adequately describes the necessary unloading operations. The unloading procedure descriptions should present the activities sequentially in the anticipated order of performance, including those key prerequisite and preparation tasks that should be accomplished before storage container unloading. Where applicable, verify that the procedure guidance in the SAR ensures that any fluids, such as shield or borated water, fill their respective cavities according to design specifications. Additionally, for storage containers that require borated water to maintain subcriticality, ensure that the procedure guidance in the SAR includes verification that the water to be used for container reflood meets the minimum soluble boron content required by the technical specifications. Verify that the operations descriptions in the SAR incorporate the applicable operating controls and limits described in the chapter of the SAR on technical specifications and operation controls and limits.
Damaged Fuel Ensure that the SAR includes appropriate additional measures for the potential presence of damaged fuel. Procedures should be designed to maximize worker protection from unanticipated radiation exposures or contaminants from damaged fuel in accordance with ALARA principles and, to the maximum extent possible, to prevent any uncontrolled releases to the environment.
The following points outline the relevant safety concerns and one acceptable approach to address damaged fuel contingencies in unloading:
The procedure descriptions should provide for fuel unloading under normal conditions.
The unloading process should ensure that the fuel can be safely unloaded with regard to structural, criticality, thermal, and radiation protection considerations. This includes the provision for safe maintenance of the fuel and storage container while any additional measures needed to address suspected damaged fuel are planned and implemented.
The unloading process should reflect the potential for damaged fuel and changing radiological conditions.
The process should include measures to check for and detect damaged fuel conditions (such as cask (or canister) atmosphere samples) before opening the storage container.
(Note that fuel oxidation resulting from exposure to air at temperatures typical for dry storage is a known form of fuel degradation. Therefore, the presence of air in a storage container designed to maintain an inert atmosphere indicates that the fuel may be degraded. The detection of fission gases is another indicator that the fuel may be degraded.)
The process may establish sample result thresholds above which damaged fuel is suspected. Other technically sound methods may be used to check for potential air
11-17 leakage paths. Such methods may include designs that monitor storage container internal pressure or seal integrity and alert the licensee to a problem before oxidation could occur. However, this method may not address detection of potential fuel degradation resulting from other mechanisms (such as a storage container drop accident).
If the sample indicates normal conditions, the normal unloading process should be followed.
If damaged fuel is suspected or found, the procedure description should stipulate that additional measures, appropriate for the specific conditions that include the canning of the damaged fuel, are to be planned, reviewed, and approved by the designated approval authority and implemented to minimize exposures to workers and radiological releases to the environment. These additional measures may include provision of filters, respiratory protection, and other methods to control releases and exposures in accordance with ALARA.
Cooling, Venting, and Reflooding Verify that the SAR describes applicable operational measures to control storage container cooling, venting, and reflooding (when appropriate). Verify that these measures are consistent with the results of the structural, materials, and thermal evaluations in the SAR, respectively.
Storage container cooling, venting, and reflooding should not result in damage to the fuel.
Operational measures may include external cooling of the storage container for initial temperature reduction, restricting reflood flow rates to control and limit internal pressure from steam formation, and limiting cooldown rates.
Devote special attention to reviews in this area since analyses of existing designs have predicted fuel temperatures during storage and transfer in excess of 260 °C (500 °F) for design-basis heat loads. Operational controls may be required to address the following potential effects during a cooldown and reflood evolution:
Storage container pressurization may occur as a result of steam formation as reflood water contacts hot surfaces.
Excessive cooling rates may cause fuel cladding and fuel rod component damage and release of radioactive material as a result of stress (e.g., thermal, internal pressure) beyond material strengths (see Sections 8.5.15.2.3, Drying Adequacy, and 8.5.15.2.4, Maximum (Peak) Cladding Temperature, of this SRP).
Excessive cooling rates may induce thermal stress that causes gross deformation of the fuel assembly components and subsequent binding with the basket.
Storage container supply and vent line failures from inadequate design for pressure and temperature could result in radiological exposures and personnel hazards (e.g., steam burns).
Fuel Crud Verify that the procedure descriptions in the SAR include contingencies for protection from fuel crud particulate material. Appendix E to ANSI/ANS 57.9 provides a short discussion of crud with
11-18 respect to dry transfer systems. Verify that the unloading procedures include an alert to operations personnel to wait until any loose particles have settled and to slowly move the fuel assemblies to minimize crud dispersion in the SNF pool. Experience with wet unloading of boiling-water reactor fuel after transport has involved handling significant amounts of crud. This fine crud, which includes cobalt-60 and iron-55, will remain suspended in water or air for extended periods. The reflood process, during unloading of boiling-water reactor fuel, has the potential to disperse crud into the fuel transfer pool and the pool area atmosphere, thereby creating airborne exposure and personnel contamination hazards. By contrast, no significant crud dispersal problems have been observed in handling pressurized-water reactor fuel because of differences in the characteristics of crud on this type of fuel.
ALARA Verify that the procedure descriptions in the SAR incorporate ALARA principles and practices.
These may include provisions to perform radiological surveys, implement exposure and contamination control measures, or use temporary shielding and inclusion of caution statements related to specific actions that could change radiological conditions.
Offsite Release Where applicable, verify that the SAR describes methods to minimize offsite releases. These methods may include filtered ventilation, decontamination of the storage containers, temporary containments, and the methods described in Section 11.5.2.5 above. The procedures should also provide for minimizing generation of radioactive waste.
11.5.5 Repair and Maintenance (SL)
A concern for review of any storage container repair capability incorporated into the DSF is that the applicant recognizes the need for inspection of loaded containers and for container decontamination. This need would apply to the storage containers used at the site as well as any loaded transportation packages received at the site. If the licensee will provide a repair capability on site for the repair of storage containers and related SSCs (e.g., overpacks and onsite transfer casks) and transport packages, verify that the SAR describes the skills and equipment necessary for performing such repairs. Section 12.5.4, Maintenance Program, of this SRP provides guidance useful for the evaluation of maintenance and repair operations.
11.5.6 Other Operating Systems (SL)
For other systems that are also considered important to safety, review the description of the locations of the various systems in relationship to their functional objectives. Verify that the applicant has described provisions for coping with unscheduled occurrences so that a single failure within one of the auxiliary systems will not result in a release of radioactive material or unanalyzed conditions that may affect any safety functions, such as nuclear criticality safety, of the DSS or DSF SSCs. Evaluate the systems to ensure that the design includes performance under normal operating loads, off-normal operating loads, loading situations resulting from primary failure and/or accident conditions, and loading situations required for the safety of a shutdown operation. If a system requires a technical specification, verify that the SAR includes the required technical specification, and ensure that it is part of the license.
11-19 11.5.7 Operation Support Systems (SL)
Review the descriptions of the I&C systems in the SAR and determine whether the applicants definition of their function is adequate. Ensure that, for SSCs important to safety, the SAR describes all major components, operating characteristics, locations of sensors and alarms, threshold levels for I&C that produce alarms, automatic and manual control actions to be triggered, and safety criteria.
Consider the projected accident and off-normal events (addressed in SRP Chapter 16, Accident Analysis Evaluation) and the roles that the I&C systems have in avoiding or mitigating significant radiological consequences of those events. Verify that the applicant has considered the redundancy required to ensure safe operation or safe curtailment of operations under accident conditions. Verify that the SAR reflects that spare or alternative instrumentation, if provided, has been designed to ensure safe functioning.
Ensure that the applicant has proposed technical specifications that include reliance on an I&C system performance as outlined in RG 3.62 and RG 3.48.
11.5.8 Control Room and Control Area (SL)
Review the control room and control area functions, equipment, I&C links, and staffing for consistency and appropriateness for the intended functional control and safety roles. Information on these different aspects of the control room or control area, as applicable, may be at various locations within the SAR.
Ensure that the SAR includes an explanation for an omission of a control room, monitoring room, control area, or monitoring area, as applicable. Explanations might include, but not be limited to, a description of functions and procedures (flowcharts and narrative descriptions) that provide for performance without the need for a centralized control room, the acceptability of accident and off-normal event and condition analyses that show acceptable levels of maximum response and safety without use of a control room, and the desire that damage avoidance and mitigation be based on passive measures to the extent feasible.
11.5.9 Analytical Sampling (SL)
Verify that the types of samples and rates of sampling are appropriate for the conditions being monitored. Ensure that the SAR includes provisions for obtaining samples during off-normal conditions to ensure that prescribed limits have not been exceeded. Confirm that the SAR describes the facilities and equipment that will be available to perform the analyses. Ensure the SAR also describes disposition of laboratory wastes.
Compare the proposed analytical sampling operations with those of existing similar facilities as documented in the final SARs for licensed DSFs. Determine whether the proposed analytical sampling operations are reasonable and the descriptions of the operations, facilities, and equipment are adequate given this comparison.
11-20 11.5.10 Fire and Explosion Protection (SL)
General Consideration (SL)
Depending on the design, magnitude, scope, and fire hazards of a proposed DSF, the applicant may have to institute a fire protection program (FPP) to satisfy the requirements of 10 CFR 72.122(c). Ensure that the applicant performed a fire and explosives hazards analysis of the facility and, if warranted, instituted an FPP. The applicant may use the following guidance:
RG 1.189, Fire Protection for Nuclear Power Plants, as it relates to the design provisions given to implement the FPP RG 1.78, Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release, as it relates to habitable areas, such as the control room and to the use of specific fire extinguishing agents the NRC technical position on fire protection for fuel cycle facilities Spent Nuclear Fuel Storage Containers (SL)
The DSF may use DSSs approved under Subpart L, Approval of Spent Fuel Storage Casks, of 10 CFR Part 72, provided, in part, that the applicant satisfies the fire requirements identified in the CoC, if any, and 10 CFR 72.122(c).
Verify that the SAR indicates that the DSS materials, such as protective coatings, are compatible with water used in the DSS cavity so as to preclude or minimize the potential for combustible gas generation. For background, refer to NRC Bulletin 96-04, Chemical, Galvanic, or Other Reactions in Spent Fuel Storage and Transportation Casks, dated July 5, 1996.
Guidance for a Fire Protection Program (SL)
Verify that waste confinement systems important to safety have adequate fire and explosive protection. Specifically, verify that an FPP provides assurance that a fire will not impact the ability of SSCs important to safety to continue to effectively perform their safety and design functions in accordance with the general design criteria in 10 CFR 72.122(c). This includes adverse effects from both the operation and the failure of the fire suppression system. A defense-in-depth approach should achieve balance among prevention, detection, containment, and suppression of fires. Confirm that the SAR indicates that there is a fire protection policy for the protection of SSCs important to safety at each facility and for the procedures, equipment, and personnel required to implement the program at the site. Ensure that the FPP consists of fire detection and extinguishing systems and equipment, administrative controls and procedures, and trained personnel.
Portions of the review procedures of Section 9.5.1.1, Fire Protection Program, of NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, and the guidelines of Chapter 7 Fire Safety of NUREG-1520, Standard Review Plan for Fuel Cycle Facilities License Applications, Revision 2, may apply to the MRS or ISFSI contingent on the design of the installation and associated fire hazards. Many of the national codes and standards cited in these NRC guidance documents, in particular, the codes and standards of the National Fire Protection Association, could apply to the ISFSI or MRS.
11-21 Review the SAR to determine that the appropriate levels of management and trained, experienced personnel are responsible for the design and implementation of the FPP in accordance with RG 1.189.
Review the SARs analysis of the fire potential in facility areas important to safety and the hazard of fires to these areas to determine that the proposed FPP is able to ensure that DSF SSCs important to safety will continue to effectively perform their safety and design functions in the event of a fire.
Review the elevated temperatures that may be of concern because of their effects on strength, heat treatment, durability, other properties, or change of state. A small amount of exterior concrete spalling may result from a fire or other high-temperature condition or application of fire, water, or rain on heated surfaces. Spalling from temperature gradients typically is considered to have minor (at most) structural significance, but such condition could partially block ventilation passages, depending on the design.
Evaluate the FPP piping and instrumentation diagrams (P&IDs) and facility layout drawings to verify that facility arrangement, buildings, and structural and compartment features that affect the methods used for fire protection, fire control, and control of hazards are acceptable for the protection of safety-related equipment.
Determine that design criteria and bases for the detection and suppression systems for smoke, heat, and flame control are in accordance with the fire protection guidance in NUREG-0800, Section 9.5.1.1, Fire Protection Program and NUREG-1520, Chapter 7, Fire Safety, and provide adequate protection for SSCs important to safety. Determine whether fire protection support systems, such as emergency lighting and communication systems, floor drain systems, and ventilation and exhaust systems, are designed to operate, consistent with this objective.
Verify the results of an FPP failure modes and effect analysis to assure that the entire fire protection system for one safety-related area cannot be impaired by a single failure.
Verify that the applicants technical specifications for fire protection, specifically the limiting conditions for operation and surveillance requirements of the technical specifications, are in agreement with the requirements developed as a result of the staff review. RG 1.189 provides guidance for fire detection and suppression as well as the fire protection water system.
Confirm that the control room or control area ventilation system P&IDs show monitors located in the system intakes that are capable of detecting radiation, smoke, and toxic chemicals. Ensure that the monitors actuate alarms in the control room. Confirm that the P&IDs show provisions for isolation of the control room upon smoke detection at the air intakes. Although the isolation may be actuated manually for most cases, special cases may require automatic isolation, such as for fires resulting from aircraft crashes. Consult RG 1.189 for additional guidance.
Verify that miscellaneous areas, such as shops, warehouses, auxiliary boiler rooms, fuel oil tanks, and flammable and combustible liquid storage tanks, are located and protected so that a fire or effects of a fire, including smoke, will not adversely affect any SSCs important to safety.
Confirm that acetylene-oxygen gas cylinder storage locations are not in areas that contain or expose equipment important to safety, or the fire protection systems that serve those areas important to safety, exposing these locations to explosive hazards. The applicant should propose a permit system to use this equipment in areas of the facility that are important to safety (also see
11-22 RG 1.189). Verify that unused ion exchange resins and hazardous chemicals are not to be stored in areas that contain or expose equipment important to safety.
Verify that materials that collect and contain radioactivity, such as spent ion exchange resins, charcoal filters, and high-efficiency particulate air filters, are stored in closed metal tanks or containers that are located in areas free from ignition sources or combustibles. These materials should also be protected from exposure to fires in adjacent areas. Consideration should be given to requirements for the removal of decay heat from the radioactive materials.
Evaluation Findings The NRC reviewer should prepare evaluation findings upon satisfaction of the regulatory requirements in Section 11.4 of this SRP. If the documentation submitted with the application fully supports positive findings for each of the regulatory requirements, the statements of findings should be similar to the following:
F11.1 (SL)
[If applicable] The DSF is to be located on the same site as another facility licensed by the NRC. Potential interactions between these facilities and the DSF have been evaluated, in accordance with 10 CFR 72.24(a) and have been determined to be acceptable and pose no undue risk to any of the facilities.
F11.2 The SAR includes acceptable descriptions and discussions of the DSS or DSF operations, operating characteristics and safety considerations, in compliance with 10 CFR 72.24(b) or 10 CFR 72.234(f).
F11.3 (CoC) The [DSS designation] is compatible with [wet/dry] loading and unloading in compliance with 10 CFR 72.236(h). General procedure descriptions for these operations are summarized in Chapter(s)___of the applicants SAR.
Detailed procedures will need to be developed and evaluated on a site-specific basis.
F11.4 The DSS or DSF storage container design allows for ready retrieval of the SNF and, as applicable for a DSF, reactor-related GTCC waste, and HLW for further processing or disposal as required. The descriptions of the proposed [DSS or DSF] functions and operating systems with regard to retrieval of stored radioactive material from storage, in normal and off-normal conditions, are acceptable and comply with 10 CFR 72.122(l) and with 10 CFR 72.236(m).
F11.5 The smooth surface [or other feature] of the DSS or DSF SSCs is designed to facilitate decontamination in compliance with 10 CFR 72.126(a)(2) and 10 CFR 72.236(i). Only routine decontamination will be necessary after the storage container is removed from the SNF pool.
F11.6 (SL)
Radioactive waste expected to be generated during operations associated with the DSF will be minimized in compliance with 10 CFR 72.24(f). [Note that contaminated water from the SNF pool will be governed by the 10 CFR Part 50 or 10 CFR Part 52 license conditions for DSFs co-located with and using those facilities.]
11-23 F11.7 No significant radioactive effluents are expected to be produced during storage. [Note that any radioactive effluents generated during the storage container loading will be governed by the 10 CFR Part 50 or 10 CFR Part 52 license conditions for DSSs and for DSFs co-located with a 10 CFR Part 50 or 10 CFR Part 52 licensed facility.]
F11.8 The content of the operations descriptions in the SAR is adequate to protect health and minimize damage to life and property that is in compliance with 10 CFR 72.24(h) for a DSF or 10 CFR 72.234(f) for a DSS.
F11.9 The radiation protection chapter of this SER evaluates the operations descriptions and systems, including implementation of operational limits and restrictions to meet the applicable regulatory requirements in 10 CFR Part 20 and in 10 CFR Part 72 (i.e., 10 CFR 72.104 and 10 CFR 72.126) for a DSF or, for a DSS, to facilitate compliance with these requirements by licensees using the DSS and to meet 10 CFR 72.236(d). For a DSS, a licensee using the DSS may also establish additional restrictions for use of the DSS its site.
F11.10 (SL)
[One of the following, as appropriate]
The design of the [DSF designation] provides for an acceptable [control room/control area] as part of the facilities to be built, in compliance with 10 CFR 72.122(j).
- OR -
The operating procedures and schedule of operations for the [DSF designation] acceptably provide for control during storage operations to be accomplished from the security, monitoring, or surveillance office facility, as appropriate, and for control during loading, transfer, and unloading operations from temporary control facilities, and the design includes acceptable provisions for such facilities. This is considered to comply with 10 CFR 72.122(j).
- OR -
The [DSF designation] is to be located on a site with existing facilities suitable and available for control of [DSF designation] operations under off-normal or accident conditions, and their use will not interfere with other operations on the site important to safety, in compliance with 10 CFR 72.40(a)(3) and 10 CFR 72.122(j).
F11.11 (SL)
The proposed [DSF designation] facilities include the following utility service systems: [identify]. [If appropriate] The following utility service systems are important to safety: [identify]. The [DSF designation] design provides for redundant systems to the extent necessary to maintain, with adequate capacity, the ability to perform safety functions, assuming a single failure, in compliance with 10 CFR 72.122(k)(1).
11-24 F11.12 (SL)
The proposed design of the [DSF designation] emergency utility services acceptably permits testing of the functional operability and capacity of each system and permits operation of associated safety systems, in compliance with 10 CFR 72.122(k)(2).
F11.13 (SL)
The proposed design of the [DSF designation] includes the following systems and subsystems that require continuous electric power to permit continued functioning of all systems essential to safe storage: [identify].
The design of the [DSF designation] acceptably provides for timely emergency power for these systems and subsystems, in compliance with 10 CFR 72.122(k)(3).
F11.14 The design and procedures for the DSF provide acceptable capability to test and monitor components important to safety, in compliance with 10 CFR 72.128(a)(1), for DSFs, and 10 CFR 72.234(f), for CoCs.
For a DSF only, if the design of the SNF storage system to be used at the DSF has been previously certified under 10 CFR Part 72, Subpart L, the following evaluation finding statement would also be appropriate:
The proposed DSF uses a SNF storage system that has been previously certified by the NRC.
F11.15 (SL)
The staff concludes that the site-specific fire and explosion hazards are acceptable and that the fire protection program meets the requirements in 10 CFR 72.122(c). This conclusion is based on the applicant meeting the guidelines in RG 1.189, Fire Protection for Nuclear Power Plants, as well as the applicable industry standards. In meeting these guidelines, the applicant has provided an acceptable basis for the [ISFSI/MRS]
design and location of safety-related structures and systems to minimize the probability and effect of fires and explosions; has used noncombustible and heat-resistant materials whenever practical; and has provided fire detection and firefighting systems of appropriate capacity and capability to minimize adverse effects of fire on SSCs important to safety.
The reviewer should provide a summary statement similar to the following:
The staff concludes that the operations descriptions, including procedures and guidance, for the operation of the [DSS or DSF] are in compliance with 10 CFR Part 72 and that the applicable acceptance criteria have been satisfied.
The evaluation of the operations descriptions provided in the SAR offers reasonable assurance that the DSS or DSF will enable the safe storage of SNF and, as applicable for DSFs, reactor-related GTCC waste and HLW. This finding is based on a review that considered the regulations, appropriate regulatory guides, applicable codes and standards, and accepted practices.
11-25 References 10 CFR Part 20, Standards for Protection Against Radiation.
10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities.
10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants.
10 CFR Part 71, Packaging and Transportation of Radioactive Material.
10 CFR Part 72, Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater than Class C Waste.
American National Standards Institute (ANSI) N14.5, Radioactive MaterialsLeakage Tests on Packages for Shipment.
ANSI/American Nuclear Society (ANS) 57.9, Design Criteria for an Independent Spent Fuel Storage Installation (Dry Storage Type).
ANSI/ANS 3.2, Managerial, Administrative, and Quality Assurance Controls for the Operational Phase of Nuclear Power Plants.
American Society of Mechanical Engineers, Boiler and Pressure Vessel Code Section III, Rules for Construction of Nuclear Facility Components Knoll, R.W. and E.R. Gilbert, Evaluation of Cover Gas Impurities and Their Effects on the Dry Storage of LWR Spent Fuel, PNL-6365, DE88 003983, Pacific Northwest National Laboratory, November 1987.
NRC Bulletin 96-04, Chemical, Galvanic, or Other Reactions in Spent Fuel Storage and Transportation Casks, July 5, 1996.
NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition.
NUREG-1520, Standard Review Plan for Fuel Cycle Facilities License Applications, Revision 2, June 2015.
NUREG/CR-4775, Guide for Preparing Operating Procedures for Shipping Packages, UCID-20820, Lawrence Livermore National Laboratory, December 1988.
Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation).
Regulatory Guide 1.78, Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release.
Regulatory Guide 1.189, Fire Protection for Nuclear Power Plants.
Regulatory Guide 3.48, "Standard Format and Content for the Safety Analysis Report for an Independent Spent Fuel Storage Installation or Monitored Retrievable Storage Installation (Dry Storage)."
11-26 Regulatory Guide 3.61, Standard Format and Content for a Topical Safety Analysis Report for a Spent Fuel Dry Storage Cask.
Regulatory Guide 3.62, Standard Format and Content for the Safety Analysis Report for Onsite Storage of Spent Fuel Storage Casks.